The reactor channel of the horizontal core of pressurized heavy water reactors experiences very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders caused by certain breaks known as critical channel breaks. In this type of accident the reactor trip is delayed causing a gross mismatch of the heat generation and heat removal in the channel, thus leading to rapid temperature rise in the affected channel. A study has been carried out to identify the phenomena and the break size leading to such a situation. Severe fuel damage is predicted in the channel.
Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR
Contributed by the Pressure Vessels and Piping Division and presented at the Pressure Vessels and Piping Conference, Seattle, Washington, July 23–27, 2000; of THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS. Manuscript received by the PVP Division, December 4, 2000; revised manuscript received December 18, 2001. Editor: S. Y. Zamrik.
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Mukhopadhyay, D., Majumdar , P., Behera , G., Gupta, S. K., and Raj, V. V. (May 1, 2002). "Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR ." ASME. J. Pressure Vessel Technol. May 2002; 124(2): 161–167. https://doi.org/10.1115/1.1463036
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